ML19256F730

Hazard Analysis Summary Rept

ML19256F730
Person / Time
Site:	University of Missouri-Rolla
Issue date:	12/11/1979
From:	MISSOURI, UNIV. OF, ROLLA, MO
To:
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ML19256F726	List: ML19256F725 ML19256F727 ML19256F729 ML19256F730 ML19256F731
References
NUDOCS 7912200465
Download: ML19256F730 (87)

Text

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I Hazard Analysis I

University of Missouri-Rolla Reactor I

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Hazard Analysis Summary Report University of Missouri - Rolla Nuclear Reactor Facility I

1628 250 I

I I

I I.

THE REACTOR, ITS SUPPORTING FACILITIES AND SITE A.

Reactor 1.

Fuel Element 1

2.

Supporting Structure 11 3.

Control Rods and Drives 12 4.

Master Control Console 16 5.

Reactor Pool 19 B.

Supporting Facilities 1.

Reactor Building 21 2.

Pool Water Supply System 21 3.

Pool Water Purification System 26 4.

Beam Hole 27 5.

Thermal Column 28 6.

Irradiation Elements 29 7.

Pneumatic Transfer Assembly 31 8.

Radioactive Waste Handling Facilities 31 9.

Ventilation 33 C.

Reactor Site 34 II.

INSURING SAFE OPERATION OF THE REACTOR A.

Reactor Control and Safety System 1.

Fission Chamber 38 2.

Startup Chamber 39 3.

Linear Power 40 4.

Log Pc rer Channel 40 5.

Safety Channels 40 6.

Remote Area Monitoring System 40 7.

Neutron Monitor 41 B.

Administrative Controls C.

Duties of Personnel 1.

Reactor Safeguards Committee 43 2.

Director of Facility 45 3.

Reactor Manager 45 4.

Senior Operators 46 1628 251 I

I

I III.

ACCIDENTS INVOLVING THE REACTOR A.

Power Excursions 48 B.

Loss of Coolant 49 I

C.

Fuel Element Failure 50 D.

Refueling Accident 50 E.

Startup Accident 52 F.

Experimental Accident 52 G.

Spillage of Radioactive Materials 52 H.

Accidents Due to Nature or Calamity 54 I.

Healing Effects 55 I

J.

Pool Surface Radiation Level 55 K.

Flooding of Irradiation Facility 55 L.

Effectiveness of Reactor Wall Shielding 56 M.

Maximum Credible Accident 56 APPENDIX I -- Geology, Soil and flydrology of Rolla Area A.

Geology and Soil B.

Ilydrology 1.

Ground Water 2.

Surface Water C.

Frequency and Magnitude of Seismic Activity APPENDIX II -- Meteorological Appraisal of Rolla Area A.

Source of Data B.

Climatological Review 1.

Surface Wind Direction 2.

Winds Aloft I

3.

Precipitation 4.

Conclusion 1628 252 I

I

I TABLES I.

Chaiacteristics of UMR Reactor 1

II.

Control Rod Parameters 14 I

III.

Impurities in Well Water 26 IV.

Population 36 V.

Population Within Given Radius 36 VI.

Safety and Control Instrumentation 41 VII.

Administrative Controls 43 I

VIII.

Reactivity Requirements 48 IX.

Earthquakes 68 X.

Annual Frequency of Wind Direction and Average Speed 77 XI.

Annual Frequency of Wind Speed 78 XII.

Average Number of Days of Precipitation Per Year 79 I

XIII.

Maximum Precipitation 80 I

I 1628 253 I

I

I FIGURES Page 1 - MTR Fuel Element 4

5 2 - Typical Reactor Core 3 - Triga 4-Rod Assembly..............................................

6 4 - Triga Fuel Rod....................................................

7 5 - Triga Instrumented Rod............................................

8 6 - Side Profile of Reactor...........................................

9 7 - End Profile of Reactor............................................

10 8 - Control Rod Drive System..........................................

17 9 - Control Console 18 20 10 - Facility Layout 11 - Reactor Building..................................................

22 llb - Facility Layout 23 12 - Re ac to r Poo l Wa te r S ys te m.........................................

24 13 - Reactor Pool Water System.........................................

25 14 - Isotope Production Element and Core Access Element................

30 32 15 - Drainage Routes 16 - Contour Intervals of Reactor Site 35 17 - Administration 44 18 - Decay Power after Infinite Irradiation 51 19 - Geologic Column...................................................

60 20 - Subsurface Contour Map............................................

61 62 21 - Geological Map of the Rolla Area 22 - Annual Frequency and Wind Direction...............................

76 I

1628 254 I

I.

Tile REACTOR, ITS SUPPORTING FACILITIES AND SITE A.

Reactor some of the more important characteristics of the reactor are tabulated in Table I.

Table I Characteristics of the University of Missouri at Rolla Reactor Type:

Swimming Pool (modified BSR-type)

Core:

lieterogeneous--uranium, aluminum, water Al/li 0 Volume Ratio:

0.7 +.05 2

Moderator:

Light Water Reflector:

Light Water and Graphite Coolant:

Light Water with free convection flow Biological Shield:

Light Water and norral concrete I

Critical Mass:

2.7 Kg U-235 for water reflector Power Level:

Up to 200 Kw Average Thermal Flux 1.6 x 10 n/cm /sec at 200 Kw with an 110 reflector 2

Temperature Reactivity Coefficient: -1 x 10 AK/K/ F I

Void Reactivity Coefficient:

-2 x 10' AK/K/cm Prompt Neutron Lifetime:

4.5 x 10 see Effective Delayed Neutron Fraction: 0.0075 Pool Water Temperature:

1350F (Maximum inlet temperature to core)

Fuel Plate Temperature (Maximum):

315 P Maximum IIcat Flux:

1.69 x 10 BTU /hr. ft Average IIeat Flux:

1.08 x 10 BTU /hr. ft 1.

Fuel Elements The reactor core is made up of Curtiss-Wright designed, MTR-type fuel elements.

The standard elements contain ten fuel bearing plates. Each plate is a sandwich consisting of a 0.020 in. think layer of aluminum-uranium oxido completely clad in a 0.020 in. thick layer of aluminum. This thickness of aluminum is sufficient to contain, under normal circumstances, all fission 1628 255 Rev. On8 g

fragments.

The uranium is full enriched in the 235 isotope. The fuel layer is approximately 2.5 in wide and contains 17 grams of U-235.

The finished plate is approximately 3 in. wide, 24 in, long and 0.060 in.

thick.

I The fuel plates are fastened into groups of ten with aluminum side plates so that the finished element has an almost square cross section measuring 3 in. by 3 in.

At one end a male guide section of circular cross section is attached and at the other end a hauling extension, bringing the over-all length of an element to 3 f t.

The guide piece is inserted into a tapered hole in the grid plate which supports the entire fuel element array or core.

The elements and grid plate are designed so that the fuel bearing plates are spaced uniformly throughout the core.

Both ends of the elements are open so that cooling water may flow between the fuel plates. The tolerances are set so that if all dimensions are off in the same direction there will be only a 20% reduction in coolant flow through any channel.

The outer surfaces of the i ner elements are cooled by water which passes through a funnel formed at the intersection of four elements and through an auxi-liary coolant hole in the grid plate.

The standard fuel elements designed by Curtiss-Wright is shown in Figure 1.

The method of inserting the element in the grid plate is indicated by Figure 2.

In addition to the standard elements, there will be rod elements, isotope production elements and core access elements.

The rod I

elaments have the central faur plates removed to accomodate the rod.

The remainder of the plates are spaced so that they have

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This reduction in coolant flow area is permitted due to the flux depression in the vicinity of the control rod channel.

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1628 261 FIGURE 5 - TRIGA INSTRUMENTED ROD I

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II UNIVERSITY cf MISSOURI at ROLLA 200 KW TRAINING REACTOR FIGURE 7 I

As of October 1, 1979 The MTR inventory of fuel elements is:

Triga inventory is:

I 22 full elements - Type 1090 5 4-Rod Clusters 4 rod elements - Type 690-C 1 Instrumented Element I left hand half element - Type 590 L 4 Rod Clusters contain 3 fuel 1 right hand half element - Type 590 R rods and 1 Control Rod Another type of fuel used in the UMRR is Triga fuel which is used along with the MTR fuel. The Triga 4-rod cluster, shown in figure 3, consists of an aluminum bottom adapter, four SS-clad fuel rods and an aluminum top handle.

An individual fuel rod contains a zirconium - hydride moderator I

homogeneously combined with 20% enriched uranium fuel. The active fuel section is 15 in. long and contains 8.5% by weight uranium.

The hydrogen-to-zirconium ratio of the fuel-moderator material is about 1,u to 1.

On either end of the active fuel section are graphite reflector pieces. The fuel and reflectors are contained in a 0.02 in. thich SS can.

Refer to figure 4.

The Triga 3-rod fuel cluster may contain an instrumented rod or a control rod guide tube along with three fuel rods.

The instrumented rod contains three thermocouples to monitor the core temperature. The instrumented rod, figure 5, as well as the fuel rods may be removed from the cluster without removing the cluster from the core.

The handles and rod locking plates on instrumented and control tube clusters are modified to accomodate the lead out tube and guide tube respectively. The neutronics of Triga fuel are well documented in various facilities in both the United I

States and overseas.

I 2.

Supporting structure 1b2b 2h4 The reactor core is supported by an aluminum tower assembly hung from a bridge which spans the pool as shown in Figures 6 and 7.

The bridge is made of structural steel, approximately 11 ft. long and 4 \\ ft. wide and it is wheel mounted on tracks located parallel to the long axis of the I

reactor pool atop the pool walls. The bridge can be moved alongs its rails for a distance of approximately 6 ft. from its normal operating position, and the reactor core when required. Stops are provided on the bridge rails to limite bridge travel withing the pool area.

The grid plate, shown in Figure 2, is made of 5 inch thick special aluminum with 54 element positions arranged in a 6 x 9 array.

The element holes pass through the grid plate to permit circulation of coolant through the core.

Auviliary coolant holes between the element holes are provided to permit coolant flow between outside plates of the fueled elements.

3.

Control Rods and Drives Three safety rods and one regulating rod are used to control the reactor.

a.

Safety Roas.

Each safety rod consists of a grooved, aluminum-coated, boron steel rod.

The nonminal dimension are:

7/8 inches thich, 2 \\ inches wide and 24 inches of active poison length. The boron content is about 1.5 to 1.7 percenter natural boron.

The safety rods serve both as a shim (coarse control) and a safety rod.

The three safety rods are magnetically coupled to their rod drive exten-sions so that in the event of power failure or receipt of a scram signal, the exciting current to the coupling magnets is interrupted and the rods I

fall freely into the core.

This feature provides a safeguard in that the normal magnet current is of such value as to limit the total weight lifted to only that required for satisfactory stable operation of the control system. A piston attached to the safety rod enters a special damping cylinder as the safety rod approaches the full insert position. The water I

forced upwards around the piston provides a hydraulic snubbing action which permits the safety rod to come to rest without damage.

1628 265 Each safety rod provides a Ak/k of approximately 3%, the exact worth varying with different core loadings. For any core loading the ganged worth of three safety rods is about 9% Ak/k.

The safety rods have a maximum rate of withdrawal of 6 in./ min.

At the most effective position (approximately 13 inches withdrawn and designated ski;n range) this speed corresponds to a rate of change in reactivity for any one rod of about 0.022% Ak/k per second.

b.

Regulating Rod.

The regulating rod is used for fine control.

It consists of a type 304 stainless steel tube having a wall thickness of 0.065 inches, a cross-sectional shape, 0.875 inches wide by 2.25 inches long with semicircular end and an effective poison length of about 24 inches. The regulating rod does not have a bottom tube end plug, and the top tube end plug contains a 3/8 inch diameter hole te permit free cir-culation of water through the tube to eliminate the danger of trapping air in the rod and producing a variable void condition.

I The regulating rod has a total worth of about.7% Ak/k depending on the core loading and a maximum withdrawal rate of 24 in./ min.

In its most effective positon, the maximum rate of change of reactivity of the regulating rod is 0.019% Ak/k per second for a water reflector.

I The regulating rod is bolted directly to the rod drive assembly instead of being connected through a magnetic coupling.

1628 266 I

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c.

General Table II I

CONTROL ROD PARAMETERS I

Safety Rods Individual Ganged Reg. Rod Ave. Reactivity Worth 3.0%

9.0%

.7%

Drive Speed (Both Directions) 6 in/ min.

6 in/ min.

24 in/ min.

Max. Rate of Reactivity Input 0.022%/sec.

0.066%/sec.

0.019%/sec.

Effective Length 2 ft.

2 ft.

2 ft.

Withdrawal at Max. Effect 13 in.

13.in.

13 in.

(Approx. )

The reactor control system is designed to permit safety rod withdrawal individually or in ganged (three rods simultaneously) operation. The ganged operation pro >'. des a simple convenient method of assuring unifrom reactor core physics ;haracteristics and facilitates routine operation of the reactor.

Individual rod control is available for training, experiments or to meet other special requirements.

I All control rod systems are equipped with console mounted electronic position indicators which measure the heights of withdrawal of their respective rods in inches. The remote position indication systems are accurate to within + 0.050 inch.

The safety rods have console mounted " insert limit", " shim range" and

" withdraw limit" lights which are actuated by micro-switches located on the rod drive meachanism.

Each safety rod magnet contains a contact-actuated micro-switch which energizes a light on the console to inform the operator that the safety rod is in contact with its magnet.

I 1628 267 I

The regulating rod has console mounted " insert limit" and " withdraw limit" switches which energize console lights.

In addition, this rod is provided with special lights to signal in which direction the regulating rod is being moved.

The regulating rod may be operated manually or automatically in conjunc-tion with the servo amplifier system to control the reactor power level at the desired operating value.

d.

Control Rod Drives. The control rods are driven by an electro-mechanical linear actuator located at the bridge.

The actuator is essentially a ball bearing type screw driven through a gear reduction unit by a low inertia servo motor.

A variable loading ratchet type drive mechanism connects the screw to the gear reduction unit.

The following description of the mechanical arrangement of the safety rod drive assembly used in this reactor outlines the design safeguards incorporated in the control rod drive system.

The arrangement of a safety rod drive assembly in normal core position is as follows:

A control element containing fueled plates and an axial hole for a control rod is inserted into the grid plate.

Attached by bolts to a special flange of control element is a stop assembly approx-imately 3 inches in height. A guide tube assembly consisting of a magnet guide tube bolted to a magnet guide extension is placed over the stop assembly and rests on the control element flange thus capturing the top end of the control element.

The top end of the magnet guide t.ube extension is fastened to the rod drive mount.

This rod drive mount is bolted to the reactor bridge.

1628 268 With this arrangement it can be seen that the lifting of a control element out of the core by movement of a safety rod is impossible without prior disassembly of the rod drive or deliberate omission of mechanical components.

In addition it is to be pointed out that a special adjustable slip clutch arrangement is incorporated between the drive motor and the linear actuator of the safety rod drive to insure that excessive loading on the safety rod drive will casue the clutch to slip thereby preventing movement of the safety rod.

Further, this special clutch is so designed that the force available to insert the safety rod is always greater than that available for withdrawal regardless of the clutch adjustment setting.

Since, for purposes of this report, a regulating rod drive assembly is essentially indentical to that of a safety rod drive assembly, the above information also applies.

In view of the design asoutlined in the above section, there appears to be no need, nor would there by any useful purpose, for procedural safeguards to prevent the inadvertent lifting of a con-trol element by a control rod when the control rod drive system is correctly installed.

See Figure 8.

4.

Master Control Console (See Figure 9)

I The control console contains all instrumentation and controls necessary for satisfactory operation of the reactor.

The console will be located in such a manner as to permit the reactor operator to observe all work being performed at the top of the pool.

Principles of human engineering and actual experience have been applied to the design of the console, the dimensions of which are approximately 87 in. long, 82 in. high, and to in.

I deep.

The prime consideration in the design of the console is centrali-I 1628 269

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zation of the safety and control rod switches, rod position indicators,

" scram" button, and flux level and period indicators and recorders to that the entire control of the reactor will be at the finger tips of the operator.

Located in accessible but less centralized positions on the console is such equipment as composite safety amplifiers, annunciator panel (22 point), radiation monitor, decade scaler with associated timer, A-1 linear amplifier, power supplies for operating the gamma compensated ion chambers, and miscellaneous pueh buttons.

5.

Reactor Pool Figure 10 illustrates the reactor pool which is essentially a rectangle approximately 19 ft. long, 9 ft. wide and 27 ft. deep and houses the reactor core and components.

Included with the pool is one beam tube and one thermal column.

Pool walls are of ordinary reinforeed concrete 18 in. thick except at the beam hole and thermal column end where the thickness is increased to 78 inches.

The increase in wall thickness extends above the floor level in a stepped arrangement at the end of the pool.

The internal concrete sides and floor of the pool have several coats of gretective vinyl paint to prevent excessive leaching of minerals from the concrete into the water.

At the opposite end of the pool from the thermal column is a fuel element storage space. This is formed by a reinforced concrete bulkhead extending 16 ft. above and 3 ft. 6 in. below the pool floor.

It is placed 2 ft.

from the main pool wall. Fuel element storage racks are installed at the bottom of this section. Then it becomes necessary to drain the reactor pool, fuel elements will be transferred to the storage rr.ck prior to draining.

The bulkhead will insure that at least 16 ft. of water covers the tops of the fuel elements at all times.

A concrete insert between 1628 272

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FIGURE 10 - UMRR FACILITY LAYOUT 1628 273 I

E00R OR 8 NA-I the bulkhead and the main pool floor insures adequate shielding to personnel working in the drained pool.

The storage section of the pool contains no drain.

B.

Supporting Facilities 1.

Reactor Building Figure 11 shows the reactor building and llB shows the building lay-out.

Insulated steel curtain walls ware used in construction of the building and a maximum amount of containment is built in.

This includes weather stripping of all doors and windows, automatic closing of all vents connected with the ventilation system when it is shut-down, and caulking of all other points where air may leak out of the building.

The main floor contains a reactor bay, control room, counting room, and office space.

At the beam hole and thermal column end of the reactor bay, the floor is dropped to provide access to the beam tube and thermal column as they emerge from the reactor pool.

All areas of the building are expected to remain free from radioactive contamination.

If the reactor bay should become contaminated, it can be closed off from the control room and office.

2.

Pool Water Supply System Supply water for the reactor pool is obtained from the University of Missouri at Rolla.

An analysis of the impurities in the well dated September 12, 1958, is shown in Table III.

1628 274

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1628 276 mute u

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Impurities in Well Water Substance Parts / Millions Calcit:m 51.5 Magnesium 30.5 I

Sodium 10 Silicon Dioxide 21.7 Free Carbon Dioxide 16.7 Sulphate, SO 28.4 4

Chlorine, C1 3.6 Bicarbonate, liCO 1

MethylorangeAlkalinity 252 In order to reduce fuel element corrosion and prevent build-up of impurities in the reactor pool water with consequent resultant neutron induced activity, the pool supply water is deionized prior, to being supplied to the pool. This is accomplished by passing water through a mixed resin bed ion exchanger (deionizer) to remove the anion and cation impurities.

The effluent water contains fewer min-erals than distilled water. A Cuno filter is installed at the dei.onizer influent to entrain undissolved solids.

3.

Pool Water Purification System While each filling and addition of water to the pool will be from a demineralized supply, there is still the possibility of impurity build-up in the water due to corrosion and 'eaching, introduction of handling implements into the water and the fact that the pool is open at the top.

In order to reduce fuel element corrosion and prevent impurity build-up, and hence build-tip of neutron induced activity, a continous purification system will be provided.

Purification will be accomplished by circulating the water through the filter and mixed bed ion exchange and routing it back to the pool.

Pool purit,- will be main-tained at a specific resistance of greater than 500,000 ohms /CC.

.y 1628 279 A schematic of the pool water system is shown on Figure 12, and the water retention system is shown on Figure 13.

4.

Beam Hole The beam hole is provided primarily to obtain a beam of neutrons which can be used for experimental purposes and to provide a wet or dry irradiation facility. The open end of the beam hole terminates at the beam room side of the pool wall and the operations required to remove or install equipment from the beam hole will be performed from the beam room.

The entire hole is lined with stainless steel.

The inner hole section wnich contains the removable beam tube has an additional lining of boral (aluminum-boron carbide-aluminum sandwich) to materially reduce activation of the stainless steel and concrete.

The removable beam tube constructed of aluminum, extends to within a fraction of an inch of the side face of the reactor core and is closed at the reactor end.

There are three arrangements which may be made here:

With the removable beam tube and beam tube extension in position this facility may be used for external beam experiments in which case a beam catcher and other suitable shielding will be required in the beam room itself.

With the beam tube extension removed from the outer beam hole section, materials for irradiation may be placed in the removable beam tube and the inner and outer shielding plugs placed in position for the I

duration of the irradiation.

With the removable beam tube removed, a blind flange may be installed between the outer and inner beam hole sections in places of the open hole spacer, in which case the reactor may be operated without use of the beam tube, 1628 280 The stainless steel lining prevents loose concrete dust which has become radioactive, from accumulating on installations within the hole and being removed with them.

The boron filling keeps the slow neutron flux to a minimum, thus preventing the production of high gamma rays by neutron capture in the concrete shield and thereby satisfying the shielding requirements.

I The outer concrete shielding plug is contained in a stainless steel s

outer casing. The inner plug is mainly of concrete encased in stainless steel. The end of the plug nearest the reactor is covered with a boral sheet to reduce activation of the plug materials. The opposite end contains a lead plug for the attenuation of gamma rays.

As an additional safeguard, there is an outer shielding door covering the opening of the beam hole. This is made watertight by means of a sealing gasket.

[

The beam tube extension is contained in and is an integral part of a stepped concrete plug.

This plug provides additional shielding besides acting as a support for the beam tube extension.

5.

Thermal Column The thermal column provides a readily accessible field of thermal neutrons for experimental purposes. Fundamentally, the thermal column consists of a 4 ft. x 4 ft. x 5 ft. graphite block extending from the reactor core to within about 3 ft. 3 in, of the outer face of the beam room pool wall.

Total shielding from the reactor core through the thermal column to the outer biological wall face is equivalent to that which would be provided by the intervening water and biological pool wall.

The thermal column irradiation facilities consist of one 1628 281 8 in. sq. and four 4 in, sq. horizontal access ports, all of which are filled with graphite plugs when not in use.

The reactor end of the thermal column is covered with a 4 in. lead shield to reduce the gamma flux in the thermal column to a minimum. The f.ont surface of the thermal column door is lined with boral.

6.

Irradiation Elements a.

Isotope Production Element. The isotope production element, Figure 14, is similar to a graphite reflector element but with a hole passing through it to permit a neutron start-up source or an irradi-ation sample to be inserted into the core.

The graphite is entirely clad in aluminum, the inner cladding forming an aluminum tube.

The element may be used as a dry irradiation facility. The top sealing plug contains a groove for an O-ring and a horizontal hole so that the plug may be secured to the complete assembly by means of an aluminum pin.

b.

Core Access Element. The core access element, also shown in Figure 14, is supplied to provide access to the active lattice of the core and is also a dry irradiation facility. The assembly is similar in shape to a fuel element and consists of a hollow aluminum can in the position usually occupied by the fuel plates. The top portion of the assembly will receive a sealing plug similar to that for the isotope production element except that it is tubular with an aluminum tube welded into its center. This aluminum tube projects upwards above pool water level and is curved under water to prevent neutron or gamma streaming out the upper portion of the pipe.

Samples for irradiation or experimental samples are dropped down the pipe on the end of a ladder.

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,sorogE ->uction uuw cme ACCESS RWENT 7.

Pneumatic Transfer Assembly (Rabbit-Tube)

The rabbit-tube is used to rapidly transmit samples to and from the reactor core.

The assembly fits into the grid plate in a manner similar to a fuel element. The system is a dual tube system, with one tube the sample tube and the other supplying the pressure differ-entail. The control system is semi-automatic, giving information relating to the sample position and irradiation time.

Provisions are made for a cadmium lining to prevent sample activation by thermal neutrons. The rabbit-tube will be used in a core configuration that will have at least one side open to a moderating medium.

8.

Radioactive Waste Handling Facilities a.

Liguid Wastes. The only liquid wastes anticipated at this time will be the pool water. Considering that the maximum operating power of the reactor is only 200 kw and that only deionized water is used in the pool, it is difficult to envision any pertinent degree -of radioactive contamination unless a fuel element has ruptured, which seems most unlikely at the specified operating power.

A pool water purification system is provided.

If it is desired to dispose of pool water which is above MFC for unrestricted areas as laid down in the Federal Regulations, it will be recirculated through the deionizers until it is below the MPC for dumpingto unrestricted areas.

I The dump line from the pool leads directly to the storm sewers in the University of Missouri at Rolla grounds.

These strom sewers lead directly to Frisco Lake, a body of water about 3 acres in area.

Frisco Lake drains into the Meremec River vis Burgher Creek, Little Dry Fork, 1628 284

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Effluent, thus collected, will be sampled and the radioactive content determined prior to dumping.

If the radioactive content is above the MPC for unrestricted use, it will either be returned to the reactor pool or absorbed in vermiculite and treated as solid waste. Action taken will depend on its specific activity and the quantity of effluent involved.

b.

Solid Wastes. Generation of radioactive solid wastes will be of a very minor nature and will be mostly stored for decay in a shielded area of the reactor building. Wastes contaminated with sufficiently long-lived material will be stored until enough has been accumulated to warrant shipping off-site to an NRC licensed disposal agency.

c.

Resins. A remote gamma dectector is mounted near the demineralizer to monitor the activity of the resins.

9.

Ventilation The ventilation system will be set up as follows:

Air flow will be from other parts of the building to the reactor bay.

When the ventilation system is shut down, all outside vents will automatically close.

It is possible to isolate the reactor bay from the remainder of the building by the closing of doors.

The Control Room, Laboratories, and Offices are air conditioned and isolated from the remainder of the building by the closing of doors.

l 1628 286 C.

Reactor Site The reactor site is on the east side of the campus of the University of Missouri at Rolla.

The school is located in Rolla, Missouri, about 75 air miles southeast of Columbia, Missouri.

Rolla is located about 100 southwest of the city of St. Louis and about 180 miles southest of Kansas City, Missouri.

In addition to being the hor;e of the University, Rolla is headquarters for the Missouri Geological Survey. A United States Bureau of Mines research division is also located in Rolla, as are important Topographic Mapping and Water Resources divisions of the United States Geo]agical Survey.

The country side near Rolla is largely hilly and rolling. Where land is cleared, the farms are largely devoted to handling beef and dairy cattle.

Many farmers also raise hogs, chickens, and turkeys.

Grape orchards are locally important east or Rolla, especially near the town of Rosati.

The reactor site (Figure 16) is pinpointed in the center of the concentric circles, immediately east of the building which now houses Metallurgical Engineering and Ceramic Engineering. The innermost circle is scaled to a radius of 100 ft.,

the next to a radius of one-eight mile, the next to a radius of one-quarter mile, the next to a radius of one-half mile and the outermost circle to a radius of one mile.

Rolla now has a population of 14,000.

Inspection of Figure 16 would indicated that about 14,000 normally live within a radius of one mile of the Reactor site.

The Univer-sity personnel, including students and staff, totals about 6,114.

During school hours about 5,000 people would normally be withing one-eighth mile of the reactor site.

During working hours about 7,000 people would be l

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,e within one-quarter mile of the site, and about 9,000 within one-half mile.

I About 25 miles southwest of Rolla, Fort Leonard Wood has about 50,000 military personnel in training. Near to the Fort, Waynesville has about 3,000 population.

I Population centers within 25 miles of Rolla, with distance and direction I

from Rolla, are tabulated in Table IV.

I Table IV Distance Distance From From Population Rolla Rolla I

Fort Leonard Wood 50,000 25 miles Southwest Salem 4,912 25 miles Southeast St. James 2,998 10 miles East Northeast I

Waynesville 3,265 25 miles Southwest Cuba 2,106 20 miles East Northeast Steelville 1,362 22 miles East Newgurg 921 10 miles Southwest I

Dixon 1,549 20 miles West Belle 1,142 23 miles North Bland 645 25 miles North I

Vienna 502 18 miles Northwest Rolla 13,823 Rolla is located near the north edge of the Ozark uplift.

Dissection of the Plateau causer the land surface to be too rough in most areas for inten-sive agricultural practice.

This accounts for the absence of large popul-ation centers with the possible exception of Fort Leonard Wood.

Approximate population distributions, in concentric rings circling the reactor site, are indicated in Table V.

I Table V I

Within 1 mile radius 14,000 Between 1 mile radius and 2.5 mile radius 1,250 Between 2.5 mile radius and 5 mile radius 1,175 I

1628 289 Between 5 mile radius and 10 mile radius 8,500 Between 10 mile radius and 25 mile radius 40,000 I

Total population within 25 mile radius 76,500 I

II.

INSURING SAFE OPERATION OF THE REACTOR A.

REACTOR CONTROL AND SAFETY SYSTEM I

When the reactor is at steady state power, the servo system may be energized to automatically maintain power level.

The servo system is interlocked so that the power level must be with + 5% of the selected power level setting before the system may be engaged, At any time the power level exceeds the 5% variation limit, the control of the reactor reverts to manual control.

The direction of motion lights serve as direction " impulse" lights during servo operation.

Two types of automatic shutdown, " rundown" and " scram", provide pretection against nuclear incidents.

The rundown action drives all control rods to their limit at full speed. This rundown feature is designed to be the first line of defense against an incipiently dangerous condition such as I

a relatively slow rate of increase in flux level above the desired operating value. A microswitch in the linear recorder initiated this action whenever the flux level exceeds 120% of the desired operating value. Other condi-tions which cause " rundown" are shown in Table VI.

The reactor " scram" action interrupts the current to the safety rod sup-porting magnets, causing the safety rods to fall into the reactor under the effect of gravity.

The scram is reserved for only the most serious situations, such as flux level exceeding 150% of the desired operating value and reactor period less than 5 seconds. Other conditions which l

1628 290 cause " scram" are shown in Table VI.

A hazardous condition warranting a scram is sensed by any one of the three ionization chambers; two chambers sensing linear level are con-nected directly to their respective safety amplifiers, and the other one is connected to the Log N System which furnishes a period siganl to the safety amplifiers.

The function of the reactor control instrumentation is to provide adequate information for determining the behavior of the reactor during any phase of operation, and to apply this information through suitable circuitry to control the reactor or to shut it down as necessary. The system used for this specific reactor incorporates (1) suitable safety devices and inter-lock circuits to reduce the neutron level under certain unsafe conditions and to actually prevent certain types of unsafe operation (see Table VI) ;

(2) pulse counting instrumentation for startup operation; (3) logarithmic neutron level measuring equipment; (4) automatic neutron flux density I

regulation and (5) various auxiliary devices for reactor operation.

I Instrumentation Systems 1.

Fission Chamber A fission chamber is used to monitor neutron flux and provide information to the console instruments during reactor startup and low level reactor operation. The chamber, encapsulated in a watertight aluminum tube assembly, is moved in and out of the core region by an electrical motor drive system. The drive system permits locating the positions of maximum chamber sensitivity during operations and retracting the chamber into a boron shield assembly when not in use.

The drive system is pro-vided with a light indication system at the console to show " Insert Limit", and " Top Half Travel", and " Withdraw T.imi t" ponit ionn of the chamber, 1628 291 g

As the fission chamber approaches its useful limit of operation, the compensated ion chambers will have developed sufficient signal to control the reactor. At this point, the fission chamber can be retracted to its " Withdrawal Limit" position withing the boron shield assembly to minimize chamber burnout at high flux levels.

The main source of neutron flux level information is obtained through ion chambers. Two compensated and two uncompensated ion chambers are used.

These neutron detectors are located adjacent to the reactor core and are equipped with coarse and fine manual adjustments for optimum positioning. They are encapsulated in watertight aluminum tube assem-blies which are mounted on the reactor bridge.

2.

Startup Chamber The startup channel is used for monitoring neutron source multiplication during critical experiments and reactor startup. The pulses from the fission chamber are fed into a preamplifier and then into a linear pulse amplifier.

The amplified signal is next fed to a decade scaler from which an accurate plot of reactor inverse multiplication is obtained and finally to a log-count-rate circuit which supplies a signal to a low-level-period meter, a log-count-rate meter, and a recorder.

From the log-count-rate meter a signal is fed to a log-count-rate recorder to obtain a permanent record of the reactor's startup behavior.

Incor-porated in the log-count-rate system is a control feature which prevents control rod movement until a minimum count rate

(> 2 counts per second)

I is attained.

This insures that the fission chamber is operating and that an adequate signal is available for an accurate plot of reactor inverse multiplication.

I 3.

Linear Power The signal from one compensated ion chamber is fed to a micromicroammeter which provides an accurate means of measuring the reactor power (neutron flux). A signal from the micromicroammeter is fed to a recorder-controller which, in conjunction with the servo-amplifier, provides automatic control of reactor power.

4.

Log Power Channel The signal from the other compensated ion chamber is fed to a log N and period amplifier which drives a period recorder and a log N recorder.

These recorders are interlocked with the reactor control and safety system as indicated in Table VI.

5.

Safety Channels Two uncompensated ion chambers serve as the detectors for two safety channels which function to immediately shut down the reactor by dropping all safety rods when the power exceeds 300 kw.

The chambers supply cur-rent signals proportional to the flux level to a preamplifier. The amplified signal is fed to a safety amplifier and magnet power supply.

If the flux level exceeds a preset value, current to the safety rod mag-nets is interrupted; thus, releasing the safety rods which drop into their full insert positions.

6.

Remote Area Monitoring System The remote area monitoring system consists of three radiation sensitive probes which are located on the reactor bridge, intermediate building level, and beam tube room respectively. The probes serve to detect radiatun in the various areas which are normally specified as " cold" areas. When the radiation in any of these areas exceeds the level set at the station indicator for that area, an adjustable trip is actuated l

1628 293 I

which causes a reactor rundown, annunciates the area effected at the console and provides local indication of this condition.

In the case of the bridge monitor, a building alarm is energized.

7.

Neutron Monitor I

A standard neutron monitoring system is position in the beam tube room to monitor the radiation level during operations when the beam tube is open.

Radiation above a set level will give an annunciation in the control room.

Table VI SAFETY AND CONTROL INSTRUMENTATION Unit Initi-Resulting Annun-Situation Dectector ating Action Action ciation 1.

Manual Scram Operator Scram Button Scram Yes 2

Period 5 sec.

Compensated Log N.

Period Scram Yes Ion Chamber Amplifier g

3 150% Full Power (2) Uncompen-Safety Amplifier Scram Yes 3

sated Ion Chamber 4

Bridge Motion Motion Switch Motion Switch Scram Yes 5

Log N and Period Log N Period Relay Scram Yes Amp.

Not Operative Amp.

6 120% Demand Compensated Linear Recorder Rundown Yes 7

Period 14 Seconds Compensated Period Recorder Rundown Yes Ion Chamber 8

Reg. Rod Insert Micro-Switch Micro-Switch Rundown Yes 9

Low CIC Voltage D.

C.

Relay D. C. Relay Rundown Yes 10 120% Full Power Compensated Log N Recorder Rundown Yes Ion Chamber I

11 liigh Radiation GM Tube RAM System Rundown Yes 12 ** Period 30 Seconds Compensated Period Recorder Rod Pro-Yes Ion Chamber hibit l

1628 294 SAFETY AND CONTROL INSTRUMENTATION (Con't)

I Unit Initi-Resulting Annun-Situation Dectector ating Action Action ciation 13 Recorder Off Relay Relay Rod Pro-Yes hibit 14 ** Log Count Rate Fission Log Count Rate Rod Pro-Yes 2 CPS Chamber System hibit 15 **(Safety Rods Below Micro-Switch Relay Rod Pro-No Shim Rangel(Reg.

hibit Rod Above Insert Limit) 16 Core Inlet Water Thermocouple Relay Rod Pro-Yes Temp. 135 F hibit 17 Interlock By-Key Switch Key Switch Yes passed 18 Effluent Pool Solo Bridge Relay Yes I

Demineralizer Conductivity High 19 High Neutron Flux Neutron Relay Yes Beam Room Detector

• Radiation detector on the reactor bridge causes building alarm.

Indicates that the situation may be key bypassed around safety circuitry.

Instrumentation The manual scram is connected to the safety amplifier magnet power. The current to the magnets is supplied from silicon diodes which are fed from the secondary winding of a power transformer.

The scram switch disconnects AC power to the primary of the transformer, thus magnet current to the electromagnet is lost.

No credible single failure of the automatic scram system could preclude failure of the system to operate. The relays which cause the automatic scram system to function are all backed up by auxiliary relays and the conditions which operate the relays are fail safe.

In any abnormal condition power to the relays is disconnected and a relay closure causes the auto-matic scram system to operate.

If the instruments are not operable it is impossible to obtain magnet power.

g 1628 295 The rundown system is a series of relays whose contacts are in parallel with the AC drive voltage for the rod drives. Again these are fail safe and a loss of current to the relays cause the drive motors to operate.

The three primary rundowns, 120% demand, 120% full power, and 15 second period come from a microswitch located in the appropriate recorders. The low compensated ion chamber rundown comes from a relay in the chamber power supply. The regulating rod on insert limit rundown comes from the insert limit switch whenever the reactor is on servo.

It is impossible for all four rods to be withdrawn simultaneously without defeating relays.

This system is interlocked such as that when either the safety rods or the regulating rod is moved, AC drive power is disconnected to the other rod drive motor (s).

B.

Administrative Controls The likeihood of an accident involving a neclear reactor can be reduced greatly by the clear definition of responsibility for the various phases of operation. The prime responsibility for the safe operation of the reactor, therefore, will be assigned to one man, the Director of the Reactor Facility.

Figure 17 shows the UMR Administration System.

I C.

Duties of Personnel 1.

Isotopes and Safety Committee Isotopes and Safety Committee shall consist of members having ex-tensive training and experience in at least one phase of reactor operations or experimental work utilizing radioactive sources.

1628 296 CHANCELL OR UMR BUSINESS RADIATION SAFETY UNIVERSITY HEALTH OFFICER OFFICER PHYSICIST l

.a _

DEAN SCHOOL OF MINES AND METALLURGY DIRECTOR NUCLEAR REACTOR MANAGER NUCLEAR REACTOR EACTOR SENIOR OPERATORS CLERICAL MAINTENANCF FLECTRONICS,

ENGINEER FIGURE 17 - UMRR ADMINISTRATION 1628 297 The responsibilities of this Committee shall be as follows:

Review all requests for reactor time which are forwarded to it.

a.

This review shall encompass only matters concerning health and safety, and shall not touch upon the technical feasibility or advisability.

b.

Approve, provisionally approve with recommendations for change in the program, or disapprove all properly submitted requests, and advise the interested parties of the outcome of the review.

Review special reports issur 3 by the Reactor Supervisor following c.

any significant malfunctions, violations, or accidents.

In addition to this review, the Committee shall either approve the I

corrective action already taken, or recommend further action.

I 2.

Director of Reactor Facility The Director will have the primary responsibility of over-seeing all reactor activities.

He shall make the final decisions relating to utilization of reactor time, feasibility of experiments, and operational procedures.

3.

Reactor Manager Assume primary responsibility for safe operation of the reactor, a.

and insure that experimental requirements do not unduly compromise safety.

b.

Supervise all reactor operating personnel.

Supervise training of new operating personnel and licensing of c.

new operators.

d.

Keep all records and logs of reactor operation up to date and in proper form.

l 1628 298 I

Manager the reactor fuel inventory, specify element rotation and e.

reprocessing schedules, keep records and issue reports concerning the full inventory.

f.

Take appropriate corrective action to prevent reoccurrence of any significant malfunction, violation, or accident in connection with reactor operations.

Issue a special report to the Reactor Safeguards Committee covering any such incident and the corrective action taken.

g.

Co-operate with the Director in scheduling and co-ordinating experi-mental programs and services irradiations.

h.

Review requests for reactor time which have been recommended by the Reactor Engineer. Approval or disapproval shall be based strictly by considerations of operational safety rather than the merits of the experiment.

4.

Reactor Engineer To plan and direct the maintenance, repairs, and modifications of the Reactor Facility. To design modifications, activations and additions to the facility.

Prepare technical specifications for equipment to meet governmental reactor licensing requirements.

Instruct and advise support staff, evaluate experiments.

5.

Senior Operator (s) a.

Accept responsibility for the safe operation of the reactor at all times during his shift except when relieved by the Reactor Manager.

b.

Make all minor decisions regarding operation of the reactor during his shift and all decisions required immediately.

c.

Remain in the reactor building at all times during his shift except when relieved; supervise routine startup, shutdown, alteration in power level, and any movement of any object in that portion of the pool in which t.'e reactor is aperating at the time.

1628 299 d.

Carry out appropriate checks of the safety circuits and supervise routine maintenance.

I Supervise the keeping of records and logs and insure that all c.

records for his shift are complete and accurate.

f.

Instruct operators and operator trainees in the theoretical and practical phases of reactor operation and maintenance.

g.

Relieve the operator at the console from time to time, and give all necessary assistance to the operator and any experimental personnel working with the reactor.

6.

Reactor Operator (s)

I A reactor operator shall be a person holding a valid NRC operator's license. Ilis duties shall include:

Manipulation of the reactor controls under the supervision of the a.

Senior Operator during startup, shutdown, alteration of power level.

b.

Manipulation of controls and surveillance of instrumentation during I

steady state operation.

c.

Remaining at the reactor console at all times during operation unless properly relieved by the shift supervisor or another operator.

d.

Keeping all such records and logs as shall be required.

e.

Advising the reactor manager of any unusual behavior on the part of the reactor and its controls, taking any necessary action to prevent damage to the reactor and protect health.

l 1628 300 I

I III.

ACCIDENTS INVOLVING THE REACTOR A.

Power Excursions I

Table VIII Reactivity Requirements I

Source Reactivity Required Negative temperature coefficient

.1%

Poisons

.4 Adequate period

.2 Compensation for loss of thermal column when out in pool

.6 Experimental requirements

.2 1.5%

Excess reactivity shall be measured for a given core loading from a clean cold critical condition. A designated core loading may include irradiation facilities such as pneumatic tube, isotope production elements, core access elements or other facilities of such nature that they become a portion of the core when installed.

The maximum excess reactivity above cold-clean-critical for any core loading shall be +1.5% AK/K except during shim rod calibration.

The reactor will operate at all time with the minimum amount of excess reactivity that will give an adequate period. Whenever the excess reacti-vity is ab.ve 0.7%, a licensed operator or senior operator must be at the controls except that an operator trainee shall be allowed to operate with an excess above 0.7% under the direct supervision of a senior operator.

All student operation is restricted to excess reactivity below 0.7%; student operation is always under the direct supervision of a licensed operator or senior operator.

1628 301 I

Due to the change in worth of the shim rod for each different core load-I ing, it is requested that the reactor may be loaded to an excess reacti-vity evaluating the individuaiworth of each shim rod (3.5%) for each diffe ent core loading.

The loading above 1.5% excess reactivity will not exceed 5 consecutive days at a time, and it will be done not more than twice a year for the purpose of shim rod calibration.

During excess core loaded periods, a senior operator shall be at the controls and the reactivity shall be the minumum required.

In addition to the built-in electronic safety interlocks, the use of operators and senior operators as described above shall provide adequate safety measures.

B.

Loss of Coolant The pool is specifically designed to preclude possibility of unintentional drainage.

It is of reinforced concrete construction, set in bed rock, to resist the most severe earthquake. Nevertheless, pool drainage must be considered because of the severity of the potential hazard.

The uncovered core would create a direct radiation hazard of about 10 rems /

min to personnel on the bridge. The problem of decay-heat removal during and after loss of coolant must be considered.

If the cladding temperature rose to the melting point (near 1220 F for aluminum) or if the cladding failed due to occluded gas pressure, radioactive fission products from a portion of the fuel in the core might be released to the building atmo-sphere.

I There is no danger of melting so long as the core heat can be conducted I

into water; that is, while the pool water surface is above the top of the core.

When the water no longer is above the core, then heat must be removed by other means; namely, by natural convection of air through and around the core, or by emergency measures.

1628 302 At least 5 Kw of internal heat can be dissipated from the core by natural convection of air alone.

In any incident that ca.n be reasonably antici-pated, the leakage of water from the pool would be relatively slow.

But even if the pool drained instantaneously, no fuel melting or dangerous release of radioactive fission products is anticipated. The loss of water moderator would shut the reactor down and the rate of internal heat gener-ation would rapidly decay to a small fraction of the safe 5 Kw limit, as indicated by Figure 18 which shows the decay power after shutdown from operation at at given power level for infinite time.

As a precaution against the loss of water, a fire hose is kept near the pool. This can be connected to a nearby fire hydrant in the event of an emergency and water added to the pool.

C.

FuE.1 Element Failure Fuel element failure may be caused by excessive hydraulic forces or excessive fuel temperature.

The maximum heat flux in the reactor will I

be about 10,000 BTU /hr-ft This is much less than the estimated burn-out flux.

Failure due to hydraulic pressure unbalanced or due to excessive temperature during controlled operation is therefore not to be expected.

D.

Refueling Accident The following safeguards are designed to prevent loading errors.

4 l

1.

All fuel handling is don,in accordance with written procedures.

2.

Loadings are pre-planned to include sequence of loading and positions of individual elements, also a loading schedule is prepared prior to commencement of loading.

3.

Loading operations are done under direct and personal supervision of qualified supervisory personnel.

1628 303 I

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I 1628 304 I

4.

Fuel handing tools are kept locked with the keys secured to prevent unauthorized movement of fuel.

5.

Loading of core will be from inside to outside.

E.

Startup Accident To prevent this, a safety check is made of the electronics and facilities prior to startup. Only when all the conditions are checked satisfactorily, may the shim rods be withdrawn past 6 inches.

I F.

Experimental Accidents 1.

Flooding Beam Port.

The worth of the beam port is so small that any change in it will have negligible effect upon the core.

2.

Dropping Fuel Element on or Beside Core.

A fuel element on the periphery of the core is worth less than 1.5%.

As indicated by previous work, the UMR Reactor can tolerate a step function addition of 1.5% reactivity.

3.

Removal of In-Core Experiments.

Any in-core experiment worth more than 0.2% will be positioned in a suberitical core and the reactor loaded by an approach to critical experiment.

The core will be unloaded to a subcritical base loading with known characteristic before the experiment is removed.

4.

Reactivity Effects of Pneumatic Tube Samples.

Reactivity ef fects from pneumatic tube samples will be limited to 0.2%.

G.

Spillage of Radioactive Materials A problem of importance in the analysis of a reactor location is the of fect I

on surrounding unrestricted areas of the spillage of radioactive materials.

This problem might arise, for example, if a highly-volatile liquid were irradiated in the reactor for the production of isotopes.

If, while it were irradiated in the reactor for the production of isotopes.

If, while 1628 305 it were being transferred from the reactor to a cask, it were dropped and its container broken, the atmosphere within the Ractor Laboratory could become contaminated; further, this atmosphere could conceivably be released to the surroundings in such fashion as to present a health h7zard in unrestricted areas.

This problem may be of importance when the material being irradiated is highly volatile, or is a solid in powdered form.

For a typical solid or liquid no special problem exists, other than the direct radiation from the sample and the problem of cleaning up contamination. Since the level of radiation will be known for each sample, adequate equipment for handling the sample will be available when the material is discharged from the reactor.

Equipment adequate for cleanup of spill will be kept available so that spills can be dealt with immediately, lessening the possibility of spreading contamination to adjacent areas.

The remainder of this section will deal with gases, highly volatile liquids, or powdered samples which mighfcause airborne activity in the event of a spill.

This problem is handled at the University of Missouri at Rolla by a com-bination of administrati/e and operational procedures.

For the normal situation, a concerted effort will be made to keep the concentration of contaminants in the atmosphere released from the Reactor Laboratory well below the limits as stated in Table II, Appendix B, 10 CFR Part 20,

" Standards for Protection Against Radiation". Among the procedures which will be followed to achieve this goal will be the double-encapsulating of materials to be exposed in the reactor in aluminum containers (for long exposures) or in sealed polyethylene containers for exposures of less 17 than 2 x 10 therTnal neutrons /sq. cm. with accompanying gamma ray and fast neutron fluxes. Only members of the reactor sta f f (or selected people working under their supervision) will be permitted to handle these 1628 306 g

capsules within the Reactor Laboratory. Further, a log book will be maintained of all material exposures. Ilowever, it is recognized that accidents can occur, and the amount of radioactivity which will be generated in any one sanple of material will be limited. Specifically, this amount of radioactivity will be limited such that, should a con-tainer be broken and its contents despersed in the air within the Reactor Laboratory, the concentrations discharges through the pans when averaged over one week will be within the maximiim concentrations of 10 CFR Part 20.

Since the ventilatory fans have a capacity of 35,000 cfm, the weekly flow for dilution is 10.0 x 10 ml.

Normal approvals will be given for concentrations considerably smaller than these, however, and samples of such size as to approach these limits must have special approvals. These approvals will consider all other activity discharged, and they will insure that the total stack discharge lies within permissible limits should the sample rupture.

II. Accidents due to Nature or Calamity.

1.

Earthquake. This is covered in Appendix I.

2.

Fire.

The Reactor and Reactor Laboratory is of relatively fireproof construction. Most ( f the material in the Laboratory is stone, concrete, water, or metal. Thus, the fire hazard is slight. The Fire Department has been briefed on radiation hazards in the Reactor Laboratory, and it can be depended upon in the event of a fire.

Further, the reactor I

will be shut down in the event of a fire.

3.

Severe Storm or Riot.

The reactor will be shut down in the event of any severe storm warning or when any type of rioting appears imminent.

4.

Power Failure. The control system is fail-safe, and the reactor will automatically scram in the event of loss of electrical power, whether such loss occurn from natural causes or actions or man.

1628 307 I

I.

Heating Effects The concrete pool wall and graphite will have small increases in tempera-ture during long runs at 200 Kw but no thermal problems are anticipated.

To guard against point boiling in the core, temperature monitors are positioned to gauge the temperature of the water before core entry and after core exit.

The inlet temperature monitor will be connected to an interlock which will give a rod withdrawal prohibit and a panel announ-ciation. When such conditions exist, the reactor will be manually shut-down or power reduced to a lower level.

I The coolant inlet temperature will be 1350F.

This should give an increase of less than 20 F in the outlet temperature and a maximum fuel temperature of less than 350 F.

J.

Pool Surface Radiation Level - N Activity Nitrogen 16 activity at the surface of the pool, directly above its core is 54 mrem /hr. with the nitrogen diffusers running the radiation level is dropped to less than 5 mrem one meter above the surface. All personnel with access to the reactor bridge are monitored. A RAM Station on the bridge sounds a building evacuation at levels above 10 mrem.

K.

Flooding of Irradiation Facility Experimentation has shown that flooding of the beam tube with water has no noticeable effect upon reactivity worth of the core.

Plooding of the isotope production element or core access element has been shown to cause a change of +0.007 AK/K if one of the preceeding is located in the center of the core or slightly less than +0.002 A K/F if located on core periphery.

I 1628 308 L.

[fectivaness of Reactor Wall Shielding The reactor is cmbedded in rock on three sides.

The lower level in the beam room is the only accessible area.

No detectable radiation is observed through the walls at any places at a power level of 10 Kw.

With all shielding in place a y activity of 3 mr/hr. The beam room is considered to be a radiation area and only monitored personnel are allowed to enter this area.

M.

Maximum Credible Accident The maximum credible accident is assumed to be the stoppage of coolant flow through a fuel element with the result that eight plates or about 4%

of the fuel inventory is melted.

Data from IDO-16790 "Spert I Destructive Test Program Safety Analysis Report" was used to investigate the possi-bility of an excursion caused by a sudden change in reactivity. A reactor such as the UMR Reactor would have probably a 1% fuel melt down if a 4-5 millisecond period excursion occurred. An insertion of about 2.5% AK/K would be required for this fast an excursion. A step insertion of 2.5% AK/K into the UMR Reactor is extremely remote.

Analysis of a MCA was done following the guide line of the EC report TID-14844, "Caulculation of Distance Factors for Power and Test Reactor Sites".

Values for the various parameters ware taken from the exis;ing hazards report or typical values from TID-14811.

The calculations were made or the basis that saturated act.ivites were reached for the iodine isotopes.

Under present operating schedules this is very unlikey to ever occur.

It is assumed that 4% of :he fuel is melted; of the fission products in this melted fuel only 5% will be released from the pool into the reactor building.

From the fission products released to the building 20s is assumed to be released to the outside l

1628 309 atmosphere.

It is further assumed that these fission products are released at ground level and carried at ground level by the air.

The following daytime conditions are used:

u = average wind velocity = 5 meters /sec I

n = stability parameter =.25 Cy = virtual diffusion coefficients in vertical plane =.4 (meters"! )

Cz = virtual diffusion coefficients in horizonal plane =.4 (meters"! )

A I = building leak rate = 2 x 10 %/sec (54.8 ft / min)

R1 = working breathing rate = 3.47 x 10~

(m /sec)

R2 = average breathing rate = 2.32 x 10~

(m /sec)

For these parameters:

Exclusion distance = 4 meters Low population zone = 40 meters The following nighttime conditions are used:

u = 3 meters /sec n =.55 Cz = 0.05 the remaining parameters are the same as daytime conditions.

For these parameters:

Exclusion distance = 30 meters Low population zone = 500 meters APPENDIX I Geology, Soil and Ilydrology of the Rolla Area A.

Geology and Soil I

Rolla is located toward the northern edge of the Ozark uplift.

The sedi-mentary rock section in the Rolla area averages about 1700 ft. in total thickness. This section consists largely of Paleozoic dolomites and mag-nesium limestones, but with some sandstone and shale members (Figure 19).

1628 310 Tha Cainbrain Lamotte formation, a basal sandstone, usually is encountered in deep wells.

The Lamotte uncomformably overlies pre-Cambrain meta-morphic and igneous rocks.

The geographical center of the Ozark uplif t lies to the southeast of Rolla. Consequently, tl. 2 regional dip in the Rolla area is toward the northwest, with a very gentle gradient of less than lo.

In places, however, sink structures, developed in the Gasconade, Roubiodoux, and Jefferson City formations (Figures 20 and 21) cause high local dips and even faulting.

The sink structures were caused by collapse of old solution channels in the carbonate rocks.

Surface exposures of sink structures at Rolla ordinarily show solidly compacted fillings of clay shale and sandstone of Pennsylvanian age.

Soils developed on surface exposures in the Rolla area are predominantly of the silty loam type.

In flood plains and channels of larger streams, such as the Dry Fork, deposits of almost pure quartz sands are locally developed.

B.

Ilydrology 1.

Ground Wate Wells furnishing water for the city of Rolla are cased for varying depths from the surface.

Danger of contamination of city water sup-plies from any Imssible escape of radioactive liquids at the reactor site seems to by very slight.

Dilution by ground water would also be a mitigating factor.

1628 311 I

I The Roubidoux sandstones and the Gasconade formation outcrop in stream channels which drain the reactor site toward the east (Figure 21).

Ir the unlikely event of an excursion, there may be a slight possi-bility of radioactive contamination of water supplies in wells near the outcrop areas.

Livestock drinking from the surface water drainage would be more direcly exposed than would the human population which depends largely on water from drilled wells.

Ground water is restricted to aquifers.

In order of decreasing importance with respect to wells bottomed in them, these are the Roubidoux, Gasconade, Potosi, Jefferson City, Eminence, and LaMotte formations illustrated in Figure 16.

a.

Roubidoux Formation. The most important water bearing formation in the area at the present time is the Roubidoux. Dolomite is the most abundant lithologic type, although locally the formation is composed largely of sandstone and chert.

The sandstone in the Roubidoux formation usually occurs in two beds separated by cherty dolomite.

In some locations one or three sandstones beds may be present.

Of the fifty-five water well logs studies, twenty-six wells bottom in the Roubidoux. These yield from one to twenty-five gallons per minute.

The depths of the Roubidoux wells range from 142 to 440 feet and average nearly 300 feet.

Most of the wells bottom in the sand-stone, but some bottom in the dolomite, usually only a few feet below the sandstone.

The static water levels in the Roubidoux wells, as recorded by the Missouri Geological Survey well logs, are highly variable from well to well.

The Roubidoux-Jefferson City contact in well number 2 in 1628 312

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L

'd Scale in Miles 10 0

10 20 30 I

GENERALIZED COIEL" JAR SEC'" ION OF BCCY.S EIPOSED IN NE ROLLA AREA SY5IEM STMBOL FCR'.'ATION NAME 'IMIC K.

LINOILGIC WARACTER Pennsylvanian Cherokee b.-100_'__q1gy, ehnle, sandstone, fornigeno o cong.

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_ Mississippian G-50' line stene. sandstonEch ert. s11 c.olite.

Devonian vin- _i one reported ocetiren e of fossilifco'is ls LL9Herron_Citz. _ A0 '

3erty dolomite,_EIn san ~dstone lei..es ]

I

i.oubidoux 3 10'_

one to three beds of as.,

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.J Ordivician

~,a scenade _ _ _ _,tlS i cherty dolomite. _ _ _ _ __

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Gunter

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u-4U' erratic sandstone.

'~'

__MmM Potosi s90' cherty dolmit.

j t

_D_erbr-Doerun 100' dol m.ite.

Lambrian

, MAS Davis 12 0' interbedded Shale, dolon% 11montone.

Bonneterre 260' dolomite.

I FIGt1RI:

.]

t section 14 lies 137 feet below the same contact in well number 1 in section 13, which is less than one-quarter mile distant. The slope of approximately 12 degrees between the two contact points is three times greater than the static water level slope.

This indicates circulation of water between the two points in the sandstone.

Other wells show greater static water slope compared to structural slope.

This indicates that the hydrologic properties of this aquelfer are not unifrom laterally.

b.

Gasconade Formation.

Second in importance as an aquifer in so far as the number of wells is concerned is the Gasconade Formation.

This formation consists mainly of cherty dolomite and varies in thickness within the area from 255 to 290 feet.

Twelve wells bottom in the Gasconade formation within the Rolla area.

Individual yields range from 8 to 34 gallons per minute. None of these wells are cased very deeply and the yields given above include water that comes from horizons above.

Some of the wells originally obtained water from the Roubidaux formation until successive dry seasons made deepening necessary. They have been deepened to their present depth.

The static water levels in the Gasconade wells do not vary as greatly as the static water levels in the Roubidoux wells.

They range from 334 to 978 feet above sea level. The static water is from 7 to 192 feet above the top of the formation, but the variation is due to the elevation differences of the static water level.

No relationship is I

indicated between static water level and structure.

1628 316 I

c.

Potosi Formation. Six of the seven wells that supply the city of Rolla and one University of Missouri at Rolla well obtain ground water from the Potosi formation. This rock unit consists of cherty dolomite 230 to 286 feet in thickness.

It is relatively flat lying with either local structure or a former erosinal surface as indicated by elevation relief of the upper and lower contacts of the formation.

Too few wells penetrate the potosi formation for a strict interpretation of its structure. Fissures and caverns are not uncommon in this formation.

/

The Potosi wells yield water at the rate of 300 to 580 gallons per minute with 20 to 130 feet of drawdown. These wells are cased to points below the Roubidoux; so total yields noted are obtained from the Gasconade, Eminence and Potosi formations.

d.

Other Aqtifers. Minor water producing formations are the Jefferson City, Em] 1ence, and LaMotte.

Production from the dolomitic Jefferson City fornation is weak and the formation is not important as a water producer in the Rolla area.

The Eminence formation consists of a cherty colomite with sandstone lenses. This formation provides water for two wells that bottom in it and possibly for wells that pass through it into deeper formations.

The Gunter sandstone which is about 30 feet thick and occurs at the top of the formation, provides water in other areas, but the Eminence wells in the area of this report bottom 70 and 85 feet below the Gunter.

This indicates that water in the formation comes from the cherty and sandy dolomite rather than from the sandstone at the top.

I 1628 317 g

2.

Surface Water I

From the reactor site surface drainage is toward the east.

Natural topography, modified by street fills and culverts conduct runof f to Frisco Lake, a body of water about 3 acres in surface area.

Frisco Lake, now a part of the Rolla Park System, was created by the damming of surface drainagU by the Frisco railroad fill.

Overflow from Frisco Lake drains eastward to the Little Dry Fork; I

then to the Dry Fork and Meramec Rivers.

I Downstream from the reactor site the first known use of this drainage for human consumption is at the St. Louis suburbs of Valley Park and Kirkwood. IIere wells are sunk into the Meramec River channel sands and gravels.

Perforated horizontal radials from these wells pick up water which is probably largely seepage from the Meremec.

I Ninety air miles from the Rolla reactor site, Valley Park is probably at least 180 miles away in terms of stream channel distance.

In the very unlikely event of an " excursion" and subsequent escape of radio-active fluid from Frisco Lake, it appears that tremendous dilution would occur before any fluid from the reactor site would reach Valley Park or Kirkwood water systems.

I The Meremec River enters the Mississippi River about 12 miles south and downstream from St. Louis with an average discharge greater than 1,000,000 gals / min.

At Eureka records over a 10 year period indicate that the maximum flow was greater than 12,000,000 gals / min. and the minimum flow 115,000 gals / min.

Downstream about 75 miles from the Meramec-Mississippi confluence, Cape Girardeau, Missouri is the first town to use the river for domectic water supplies.

Possibility of significant contamination of Cape Girardeau water supply from the l

1628 318 Rolla reactor site seems very remote.

I C.

Frequency and Magnitude of Seismic Activity Examination of the Bulletins of the Seismological Society of America for the period 1925-54, selected papers on the seismic history of Missouri, and others on the regional distribution of seismic disturbances revealed that, although the state of Missouri lies within a relatively inactive area, it contains six districts within the state that can be classed as minor seismic districts. These districts have been named the New Madrid.

St. Mary's, St. Louis, IIannibal, Springfield, and Northwestern districts.

Rolla does not lie in any of taese districts but is situated approximately I

in the center of a square formed by connecting the Springfield, St. Mary's, St. Louis, and Northwestern districts. There has been no recorded instance of an earthquake focus occurring in or adjacent to the town of Rolla in at least the last 140 years.

It seems reasonhle to assume, on a basis of its past seismic history and because it does not fall in one of the known seismic districts in Missouri, that it will most probably not be the focus for an earthquake in the near future. Considering first the seismic history of Missouri, shows that the first recorded instance of seismic activity was in 1811-12.

A series of earthquake shocks (now called the New Madrid series) occured over a period of more than one year, with some 1,874 individual shocks being reported. The affected area included S.E. Missouri, N.E. Arkansas and Western Kentucky and Tennessee.

These shocks are unequal in number, continuity, area affected, and severity by any earthquakes in the United States in historic time.

Visible surface effects covered an area of 50,000 square miles and felt motion occured in an area of one million square miles.

From Indian legends and historical data it would appear that this area has an earthquake history prior to 1811, 1628 319 I

I but nothing of that magnitude. The data in the attached table (compiled from the " Seismological Notes" in the S A Bulletins, and from a paper by Ross R.

Ileinrich on seismic activity in Missouri) lists the recorded earthquakes originating in Missouri from 1811-1954, with date, probable place or origin (or reporting point closest to focus), and intensity in terms of the Wood-Neumann scale.

It is evident from this list that Missouri is a fairly active minor seismic area, with fairly frequent minor shocks and occasional large ones. As previously mentions, lieinrichs and other investigators have divided Missouri into six seismic districts.

The New Madrid district is made of portions of five states, Missouri, Arkansas, Illinois, Kentucky, and Tennessee.

The Missouri section of the Seismic zone is made up of Pemiscot, Dunklin, Mississippi, New Madrid, Stoddard, Scott, and part of Butler Counties.

The earthquakes originating in this seismic district tend to occur along a line connecting New Madrid,

'harleston, and Caruthersville, strongly suggesting basement faulting along this line. Approximately 60% of the seismic activity in Missouri has originated in this district.

I The St. Mary's district is confined to Perry, Ste. Genevieve, St. Francois, and parts of Iron, Washington, Franklin, and Jefferson Counties, This district is on the northeastern flank of the ozark uplift and is traversed by a line of northwesterly trending faults. About 25% of the seismic activity originating in Missouri occurs here.

The remaining 15% of the seismic activity originating in Missouri in the past has been divided between the four remaining districts; St. Louis, Ilannibal, Springfield, and Northwestern. Frequency of earthquakes in any given seismic area cannot be predicted on any periodic basis. This is, indeed, a very controversial question among seismologists.

Many such attempts have been made to demonstrate periodic frequencies, but most I

1628 720 I

Table IX Occurrence arrt Intensity of Eartitquake Activity in Missouri Since New Madrid Shocks of 1911-1812

(*vlolent enou,1h to cause damage)

Date Place Intensity Remarks

• 1811-1812 New Madrid X11 See Text July 25, 1816 New Madrid ILI-IV April 11, 1818 St. 1.ouis III-IV Sept. 2, 1819 New Madrid III-IV Also felt in St. Louis Sept. 16, 1819 Cape Girardeau III-IV Nov. 9, 1820 Cape Gi rardeau (7)

I July 5, 1827 St. l.ouls IV Aug. 14, 1827 St. l.ouis Ili

• June 9, 1838 St. l.ouis V

an. 4, 1843 New Madrid IX One of the most severe in I

Missouri history Fe b. 16, 1843 S t.. l.ouis

( ?)

Mar. 26, 1846 New Madrid 11-11I Oct. e, 1857 St. l.ouis VII

• Auv. 17, 1865 New Madrid VLI J u ly d, 1872 Western Missouri III Nov. 8, 1875 Kansas City III I

Sept. 25, 1879 Gayoso 11I July 13, 1880 Gayoso

( ?)

July 20, 1882 Charleston V

I July 28, 1882 1ronten

( ?)

Sept. 27, 1882 Mexico VI Covered area 250 x 160 mi.

Oct. 14, 1882 Eastern Missouri V

I hov. 15, 1882 ht. Iouts til

.l an. 1i. 1883 New Madrid V

Dec. 5, 1883 1;ovenden Sprina,s VII reb. 15, 1884 Caledonia III I

Feb. 2 l, 1885 Carthage iII Au>,

31, 1886 l'.aste rn Misuouri 11 Effect of destructive earthquake at Cha r les t on, S. C.

I Uet. 18, 1895 New Madrid II oct. 31, 1895 Ch arles t on Vll-IX Felt as far as New Mexico lh c. 2, 1897 Kansas Cit y III June 14, 1898 New Madriil III Jaa. 24, 1902 St. l.o u i n VI Two severe shocks strongly felt in " Lead Belt" Oct. 4, 1903 St. louis V

I Nov. 4, 1903 New M.idrid VI Fe lt in 8 states Nov. 2 'e, 1903 New Madrid II-Ill

.iov. 25, 1903 New Madrid II I

Nov..' 7. 1903 New MadrIJ Ill Aug. 21, 1905 Mo.,

t ud., Ky, Tenn VI Considerable dam.v,e in St. l.ouis Fele. 23, 190b An a!>e 1 II I

Nr, o,

190b ilana llea I

'V 1628 321 I

I Date Place Intensity Remarks July 4, 1907 Bismark IV Nov. 10, 1907 St. Louis IV Nov. 12, 1908 Sedalia IV Oct. 23, 1909 Cape Cfrardeau V

Feb. 28, 1911 Kenwood Springs IV Ap r. 28, 1915 New Madrid IV I

May 21, 1916 New Madrid IV

• Apr. 9, 1917 St. Ma ry 's VI Considerable damage May 9, 1917 Ilendrickson III-IV I

. lune 9, 1917 New Madrid IV July 1, 1918 Ilannibal IV Oct. 15, 1918 New Madrid V

May 26, 1919 New Madrid

(?)

Feb. 28 1920 Springfield IV

• May 1, 1970 St. l.ouls V

No shock felt in Columbia I

Oct.

3, 1920 liarrisonville III

.ian. 9, 1929 New Madrid IV

• Mar. 22, 1922 New Madrid V

Slight damage Mar. 28, 1922 Popular Blut f III I

Nov. 26, 1922 St. Louis V

Some damage in St. l.ouis Oct. 28, 1923 New Madrid VII Dec. 31, 1923 New Madrid IV I

Mar. 2, 1924 New Madrid IV July 30 1925 Kansas City

(?)

Oct. 27, 1926 Popular niuil IV Dec. 13, 1926 Pe ruia 1II I

Feb.

L, 1927

.la ckson IV Feb. 3, 1927 Popular Blut f IV May 7, 1927 New Madrid VI Some damage I

Mar. 17, 1928 St. l.ouis 1

Apr. 15, 1928 New Madrid III May 31, 1928 New Madrid 1V I

Ft ti. 26, 1927 Arcadia IV apr. 2, 1930 Ca ru t he rsv i lle IV 9

May 28, 1910 llanni ha l IV Aug.

8, 19JO llannibal IV I

Sept. 1

!930 Pe rma IV De c.. 2J, 1930 St. Louis IV Apr. 6, 1(131 St. l.o u l et III I

luly 18, 1931 New Madrid IV lh2h j22 Aug.

9, 1931 Kansas City IV De c. 17, 193t St. I.o u i :

I1 I

Mat. 17, 19 3'l Poplar lit uf f IV iuly 13, 1931 St. Mary'.

111 Aug.

3, 19 33 St. M a ry ',

IV art.

W.,

19 T) Cape Gi ra ril. aii

(?)

I G iv, l te, 19 i i s. t ve :

1V

spt, t/,

19 14

t.

M.o ry '.

111 May

's, 19 l!.

St. Ma ry ' t.

III-IV July 2, 19 34 Pemiscot County ill I

I Date Place Intenalty Remarks

• Aug. 19, 1934 Charlesten V

Jan. 30, 1935 Pawnee III I

Feb. 16, 1936 Ilayti IV Oct. 20, 1936 New Madrid 1

Oct. 31, 1936 S. E. Missouri I

Jan. 30, 1937 Caruthersville III I

Mar. 38, 1937 Pe rryv ille III Oct. 5, 1917 New Madrid III Jan. 16, 1938 Pe rryvi lle III Mar. 16, 1938 New Madrid

(?)

Sept. 28, 1938 Malden III Apr. 15, 1937 New Madrid

( ?)

I Feb. 4, 1940 Cape G1rardeau III Dec. 27, 1942 Maplewood

(?)

.lan. 15, 1945 Little Saline Creek IV May 15, 1946 Doniphan III I

• June 29, 1947 St. Iouis V-VI Some damage Dec.

1, 1947 Lit tle Black River II-III Feb. 8, 1950 Lebanon IV I

Sept. 11, 195 3 S t. Louis

( ?)

Slight Feb.

2, 1954 Poplar Bluff IV Felt over wide area of S. E. Missouri I

I I

I 1628 323 I

I I

I I

I Earthquake Data I

Date Place Itensity 1-24-56 III 10-29-56 y

11-26-56 VI O

l-26-58 y

pp 1-06-59 III 12-25-61 Kansas City y

I 12-25-61 Kansas City y

pp 2-2-62 v1 7-14-62 III I

3-03-63 v1

/!

3-17-64 Iy 5-23-64 y

5-23-64 III 3-6-65 viburnum III 8-15-65 y

10-21-65 Taum Sauk Mountain yI O

I 2-13-66 Iy 7-21-67 v1 8-05-67 III pp I

l-20-69 III 2-06-70 II

/!

2-06-70 II 2-06-70 II 3-27-70 III 7-6-70 Flat River III 12-24-70 Iy I

6-09-72 Potosi III

!/

9-6-72 II l-12-73 De Soto Iy pp 10-9-73 IV I

4-5-74 Union II p

5-13-74 v1 8-11-74 Winona y

2-13-75 y

2-20-75 IV 6-13-75 v1 I

12-03-75 VI 1-23-76 v1 5-22-76 IV 12-13-76 y

I 1-3-77 y

// Southeast, Mo.

Along the new madrid fault.

g 1628 324 I

have proved negative.

Heinrich has estimated, however, that as an average Figure 4 earthquakes per year in Missouri (provided results are tabulated for at least a ten year period) could be expected. With considerably more confidence, it can be said that these earthquakes will be confined to the six seismic zones (focus, that is) and that 60% will occur in the New Madrid district, 25% in the St. Mary's district, and 15% will be spread through the remaining four dis:ricts.

The intensities of Missouri earthquakes has ranged from a minimum of 1 on the Wood-Neumann scale to the maximum recorded for any earthquake; however, 85% since 1811 have been of slight to moderate intensity.

Of the remaining 15% only 75% were stror.g enough to do considerable damage, and almost all of these earthquakes originated in the New Madrid district.

I From the above considerations it would seem that Rolla should be reasonably secure from the prospect of earthquake damage. The odds against the occurrence of an earthquake focus in or near Rolla and the intensity of any earthquake shocks felt in or near Rolla and the intensity of any earthquake shocks felt in Rolla from seismic activity in one of Missouri's seismic districts would not normally be expected to be in excess of IV and would probably be considerably less.

I APPENDIX II Meteorological Appraisal I

of the Rolla Area A.

Source of Data Weather observations taken from records of the Vichy Weather Station from October 1965 to June 1975 and temperature and rainfall data was extracted from these records.

I 1628 325 I

Direction and speed of winds was not available for the reactor site itself; however, complete records have been taken for a number ff years at the CAA station at Vichy, Missouri which is 13 miles north of the Rolla Site. The topography at and surrounding Vichy is quite similar to the Rolla area. The Vichy elevation is 1100 ft msl,~the same as that of Rolla. There seems no valid reason to assume that the data which has been collected at Vichy will not be adequate for the evaluation of the Rolla site.

Extensive research of meteorological data shows very little variations in data analyzed since 1931.

B.

Climatological Review The general climate of Missouri is a continental mid-western type.

The area has generally adequate rainfall without extreme variations from year to year. Temperatures.have, in general, a continental range with hot summers to general mild winters ranging from over 100 F to -4 F.

The prevailing wind across the area is South-Westerly.

More specific analysis of the individual elements, particularly those affecting diffusion -

material by the atmosphere follows.

1.

Surface Wind Direction Hourly wind observations for a 6 year period, 1970 to 1975, for the CAA Vichy Station were studied in detail.

Table IX presents the percentage frequency of wind directions and average velocity for the period 1970 to 1975 inclusive.

It is immediately evident that there is little variation of the most frequent winds from day to night, during periods of precipitation, and also when the visibility is low.

These figures show that, on the average, the distribution of wind directions will be about the same regardless of the type of weather that is occurring. A detailed examination of the seasonal variations show that this holds true for all four seasons.

The only major l

1628 326 variatior, with season is that the west to northwest winds are more I

frequent during the winter as would be expected and that the highest wind velocities occur during the spring.

Figure 22 shows the remarkably constant prevailing wind directions with various wind conditions somewhat more graphcially than does the table.

Major flow is from the SSW quadrant regardless of the weather conditions occurring at the time.

I Ilighest wind speeds generally flow from the NW quadrant.

The maximum wind speed observed for this period of record was 60 mph.

It is not probable that rare wind gusts might reach as high as 85 mph.

The data on winds occurring with precipitation was included in order that one might consider the effect of washout of potential airborne contaminants. The wind frequency during periods of low visibility was included as a method of estimating the wind direction during periods of atmospheric stability. Since these do not differ markedly from the day or night wind frequencies, no special consideration of variation in weather conditions seems necessary in considering the transport of pollutants by the wind.

Another point of uniformity that can be noticed in the wind at the area is the distribution of wind speeds with various weather condi-tions. Table XI illustrates the annual frequency of various wind speed classes.

It is noted that by far the largest proportion of the winds are between 4 and 12 mph averaging over 50% in all circumstances.

The second largest occurrence is in the 13 to 24 mph category.

I l

1628 327 I

2.

Winds Aloft The Winds Aloft Summary for the St. Louis, Missouri are was examined.

St. Louis is one of the nearest stations to Rolla which takes upper wind observations. The general flow of air is from the west with most frequent flow from the west-northwest quadrant.

Velocities increase steadily as the elevation above the surface increases.

I 3.

Precipitation Climatological observations for the University of Missouri at Rolla site were examined for the years 1970 to June 1975. Average annual precipitation for this period was 39 in. per year. The minimum annual precipitation was 22.5 in.

The period with most precipitation is generally April through August and the least amounts are recorded in December and January. The range of average precipitation is from about 1 inches per month at minimum periods to around 4 inches per month at the time of the rainy season. Table XII shows the average number of days with precipitation equal to or greater than certain specified amounts.

From this table, it can be seen that precipitation amounts equal to or greater than a tenth of an inch will occur about 20% of the days in a year.

IIeavy amounts of half an inch are less frequent.

It should be noted, however, that precipitation is extremely variable.

This is borne out by the range of precipitation occurrence which is presented in Part B of Table XII.

The central Missouri area, including the town of Rolla, is subject to storms producing heavy precipitation. These storms may occur in any season of the year but high intensity short duration rainfall can be expected with consider-able frequency during the spring and summer months with the passage of thunder storms over the area.

Table XIII is a listing of the maximum precipitation recorded during the period between 1970 and 1975.

1628 328 I

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ANNUAL FREQUENCY AND I

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I Table XILL Maximum Precipitation Duration (lioural Amount Date 1/2 1.88 July 1957 1

1.41 July 1952 2

3.00 April 1953 3 1/2 3.09 August 1955 4

4.96 June 1958 6

3.15 December 1957 24 3.91 July 1951 48 5.04 June 1958 I

9-I I

I I

1628 333 I

I The LaMotte formation throughout that area occurs at a depth of more than 1600 feet.

Its thickness is unknown, but may range from 250 to 500 feet based on the data outside the area.

It is considered a poor producer of water, but one known well yields about 250 gallons per minute from it in the Rolla area.

The Elvins group and the Bonneterre dolomite are non-producers of ground water in the area.

The former,

.de up of the Derby-Doerun and Avis formations, consists of beds of shale, limestone, and non-cherty dolomite. The thickness of the Elvins groups is about 260 feet thick.

I e.

Summary and Conclusions.

Six aquifers are known beneath the Rolla These supply the city and immediate area with water.

In de-area.

creasing importance, on the basis of the number of well bottomed in them, the producing strata are:

the sandstones of the Roubidoux formation, the fractured cherty dolomites of the Gasconade, Potosi, and Eminence formations, the sandy and cherty dolomites of the Jefferson City formation and the sandstone of the LaMotte formation.

The Potosi wells supply the city of Rolla and the University of Missouri at Rolla at the rate of one-half to one and one-half millions of gallons of water per day.

Figures for other formations are not available and part of the supply is from horizons above the Potos.t formation, but Potosi production probably is greater than production from other aquifers.

I The lens-like character of the sandstones and lateral cPange in lithology of the Roubidoux formation greatly influenced t he yield I

1628 334 A small proportion of the wintertime precipitation will be recorded as snow.

It can be expected that only about 26 inches of snowfall will be recorded each winter and this can be expected to melt off and not accumulate. Ileavy snowfalls are uncommon. The maximum snowfall recorded during ary one 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period was 19.5 inches on November 5, 1951.

4.

Extensive research into the weather records from 1972-1978 shows very little changes in the weather patterns presented in the original llazard Analysis. During the years covered in the latest update of climatological records a variance of less than.17% average of the established norms since 1950. The data reviewed wau a certified copy of records from the National Climatic Center, Asheville, North Carolina.

I 1628 335 I

I